ML17034A084

Vogtle Electric Generating Plant, Units 1 and 2 - Request for Additional Information (CAC Nos. ME9555 and ME9556)

ML17034A084
Person / Time
Site:	Vogtle
Issue date:	03/07/2017
From:	Miller G E Plant Licensing Branch II
To:	Pierce C R Southern Nuclear Operating Co
Miller G, NRR/DORL/LSPB, 415-2481
References
CAC ME9555, CAC ME9556
Download: ML17034A084 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 7, 2017 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc. Post Office Box 1295, Bin -038 Birmingham, AL 35201-1295

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 -REQUEST FOR ADDITIONAL INFORMATION (CAC NOS. ME9555 AND ME9556)

Dear Mr. Pierce:

By letters dated September 13, 2012, as supplemented by letters dated August 2, 2013, July 17, 2014, November 11, 2014, December 12, 2014, March 16, 2015, and May 5, 2015, February 17, 2016, April 18, 2016, and July 13, 2016, Southern Nuclear Operating Company, Inc. (SNC), submitted a license amendment request to modify the Vogtle Electric Generating Plant Technical Specifications requirements to permit the use of Risk Informed Completion Times in accordance with Nuclear Energy Institute (NEI) Report NEl-06-09, Revision 0, Risk Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS). The U.S. Nuclear Regulatory Commission (NRC) staff finds that additional information is needed as set forth in the Enclosure.

The NRC staff has held public meetings with SNC to discuss these RAls on January 26, 2015, and February 2, 7, and 28, 2017.1 A portion of the RAls were provided in final form via electronic correspondence on February 3, 2017.2 This letter transmits the remaining RAls {RAI 6, 10, and 11). Docket Nos. 50-424 and 50-425

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv Sin;r5 J (_3b .Jf1"'1 G. Edward Miller, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 1 The draft RAls discussed at the meetings is available under Agencywide Documents Access Management System Accession No. ML 17027A018 and ML 17058A127.

2 The first set of final RAls are available under ADAMS Accession No. ML 17037A175.

REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SOUTHERN NUCLEAR OPERATING COMPANY. INC. VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425 By letter dated September 13, 2012, as supplemented by letters dated August 2, 2013, July 17, 2014, November 11, 2014, December 12, 2014, March 16, 2015, and May 5, 2015, February 17, 2016, April 18, 2016, and July 13, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 12258A055, ML 13217 A072, ML14198A574, ML14315A051, ML14346A643, ML15075A479, ML15125A446, ML 16048A096, ML 16109A338, and ML 16195A503, respectively), Southern Nuclear Company, Inc. (SNC), proposed changes to the Technical Specifications (TSs) for the Vogtle Electric Generating Plant (VEGP or Vogtle). The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times (RICTs) in accordance with Topical Report (TR) Nuclear Energy Institute (NEI) 06-09, Revision 0-A, Risk Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines.

The U.S. Nuclear Regulatory Commission (NRC) staff has identified the following information needs associated with your amendment request. By electronic correspondence dated January 24, 2017 (ADAMS Accession No. ML 17027 A018), the NRC staff provided a draft request for additional information (RAI) to Southern Nuclear Operating Company, Inc. (SNC), for discussion at a public meeting. Subsequently, at public meetings dated January 26 and February 2, 2017, the NRC staff discussed these information needs with SNC. By electronic correspondence dated February 3, 2017 (ADAMS Accession No. ML 17037A175), the NRC staff provided an updated RAI that addressed issues discussed at the meetings.

Another public meeting was held on February 7, 2017, to continue discussions.

As a result of these discussions, the NRC staff has further modified RAI questions 6, 10, and 11 to provide additional clarity and regulatory basis. RAls 6 and 11 have been modified to more clearly convey what information a complete response would contain. RAI 10 was originally written as a broad question applied to multiple TS actions, however, it has been modified to more specific individual questions for each specification.

The other questions from the February 3, 2017, correspondence remain valid. RAl6: According to Section A-1.3.2.1 of Appendix A of Regulatory Guide (RG) 1.177, when a component fails, the common cause failure (CCF) probability for the remaining redundant components should be increased to represent the conditional failure probability due to CCF of these components, in order to account for the possibility that the first failure was caused by a CCF mechanism.

When a component fails, the calculation of the plant risk, assuming that there is no increase in CCF potential in the redundant components before any extent of condition evaluation is completed, could lead to a non-conservative extended completion time calculation, as illustrated by inclusion of the guidance in Appendix A of RG 1.177. Much of the discussion in Appendix A describes how configuration specific risk calculations should be performed.

Enclosure In Section 3.2 of the NRG safety evaluation for NEI 06-09, the NRG staff stated that compliance with the guidance of RG 1.174 and RG 1.177, "is achieved by evaluation using a comprehensive risk analysis, which assesses the configuration-specific risk by including contributions from human errors and common cause failures." The limitations and conditions in Section 4.0 of the safety evaluation for NEI 06-09 state that: The [NRG] staff interprets TR NEI 06-09, Revision 0, as requiring consideration of [Risk Management Actions] RMAs [due to the potential for increased risks from common cause failure of similar equipment]

whenever the redundant components are considered to remain operable, but the licensee has not completed the extent of condition evaluations

[such that a CCF mechanism can be confirmed or excluded].

The requirement to consider RMAs prior to the determination that a CCF mechanism exists or does not exist was included by the NRG staff in the safety evaluation for NEI 06-09 as a measure to account for the potential that the first failure was caused by a CCF mechanism.

However no exception to the RG 1.177 guidance was taken in the SE for NEI 06-09 for the calculation of the RICT with regards to the quantification of CCF before a CCF can be confirmed or excluded.

Please confirm and describe how that treatment of CCF, in the case of an emergent failure, either meets the guidance in RG 1.177 or meets the intent of this guidance when quantifying a RICT. Addressing CCF in this case could adjust the RICT calculation to numerically account for the increased possibility of a common cause. Alternatively, prior to exceeding the front stop, implement RMAs that are not credited in the RICT calculation sufficient to ensure that any safety function would not be lost if a CCF condition did exist and the remaining train failed to function upon demand. Either option would need to remain in effect until the possibility of CCF was excluded at which point, a new RICT could be calculated or appropriate RMAs reconsidered.

RAI 10: The Commission's Policy on Probabilistic Risk Assessment

("Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities," dated August 16, 1995) identifies five key safety principles required for risk-informed decision-making applied to changes to TSs as delineated in Regulatory Guides 1.177 and 1.17 4. They are:

• The proposed change meets current regulations;

• The proposed change is consistent with defense-in-depth philosophy;

• The proposed change maintains sufficient safety margins;

• Increases in risk resulting from the proposed change are small and consistent with the Commission's Safety Goal Policy Statement; and

• The impact of the proposed change is monitored with performance measurement strategies. RAI 10.1: NEI 06-09, "Risk Informed Technical Specifications Initiative 4b: Risk Managed Technical Specifications (RMTS)," Revision 0-A, states that Risk Management Actions (RMAs) and compensatory actions for significant components should be predefined to the extent practicable in plant procedures and implemented at the earliest appropriate time in order to maintain defense-in-depth.

Moreover, the NRC staff's safety evaluation for NEI 06-09, Section 4.0, "Limitations and Conditions," (ADAMS No. ML 12286A322) states that a licensee's LAA adopting the NEI 06-09 initiative will describe the process to identify and provide compensatory measures and RMAs during extended Completion Times (CT), and provide examples of compensatory measures/RMAs.

In the LAA dated September 13, 2012, Enclosure 10, "Risk Management Action Example," pages E10-3 and E10-4, the licensee provided two examples of risk management actions that are considered during a RICT for: (a) inoperable diesel generator, and (b) inoperable battery. Provide similar examples of RMAs to assure a reasonable balance of defense-in-depth is maintained for the following TS actions: Current Proposed TS Condition Description Completion Maximum Time Backstop 3.8.1 c Two required offsite circuits inoperable.

24h 30 days 3.8.1 D One required offsite circuit inoperable 12h 30 days AND One DG inoperable.

3.8.1 F One automatic load sequencer 12h 30 days inoperable.

3.8.4 c One DC electrical power source 2h 30 days inoperable for reasons other than Condition A or B. 3.8.7 A One required inverter inoperable.

24h 30 days RAI 10.2: As summarized from the Updated Final Safety Analysis Report (UFSAR), Vogtle's TS Bases state: "The AC, DC, and AC vital bus electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to engineered safety features (ESF) systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded." When the licensee enters TS Conditions 3.8.9.A, 3.8.9.B, or 3.8.9.C, these subsystems carry the potential vulnerability to single failures that will reduce protection against the exceedance of the design limits. Current Proposed TS Condition Description Completion Maximum Time Backstop 3.8.9A One or more AC electrical power 8h 30 days distribution subsystems inoperable.

3.8.9 B One or more AC vital bus electrical power 2h 30 days distribution subsystems inoperable.

3.8.9 c One or more DC electrical power 2h 30 days distribution subsystems inoperable

1) For each TS condition's lowest estimated RICT (least amount of time available, calculated beyond the front-stop):

a. Describe a scenario/plant configuration for this condition.

b. Explain how each subsystem would retain the ability to defend against vulnerabilities during this scenario (e.g., examples of RMAs to assure a reasonable balance of defense-in-depth is maintained for this TS condition).

2) For each TS condition's highest estimated RICT (most risk significant component(s) that would result in a calculation close to the 30-day back-stop, without Probabilistic Risk Assessment (PRA) functional consideration):

RAI 10.3: a. Describe a scenario/plant configuration for this condition.

b. Explain how each subsystem would retain the ability to defend against vulnerabilities during this scenario (e.g., examples of RMAs to assure a reasonable balance of defense-in-depth is maintained for this TS condition).

The proposed changes to the TS include Condition 3.4.11.F, Two [Pressurizer Power Operated Relief Valve -PORV] Block Valves inoperable.

The current TS require restoring one block valve to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The proposed change is to permit the option of calculating a RICT for this Required Action. Per the proposed RICT program, the RICT could be calculated to be any length of time between 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 days. The TS bases state that an OPERABLE block valve may be either open and energized, or closed and energized with the capability to be opened, since the required safety function is accomplished by manual operation.

Although typically open to allow PORV operation, the block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV that is capable of being manually cycled (e.g., as in the case of excessive PORV leakage).

A TS loss of function is considered to exist when two redundant SSCs are simultaneously inoperable.

Voluntary entry into a condition representing a TS loss of function is prohibited throughout the proposed TSs by a Note which modifies the Condition.

If emergent conditions create a TS loss of function, the RICT is limited to a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and constraints on PRA functionality are applied. The required position of the PORV block valves could be either open or closed, dependent on the condition of its associated PORV. If the block valves are not repositionable, then inoperability of the block valves could result in a loss of safety function. Please provide an explanation of how PRA functionality would be applied in this Condition, why this condition would not be considered a TS loss of function, and how it would be assured that design basis success criteria would be satisfied.

RAI 10.4: The proposed changes to the TS include Condition 3.5.1.B, One Accumulator Inoperable (for reasons other than Boron Concentration).

The current TS require restoring the accumulator to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The proposed change is to permit the option of calculating a RICT for this Required Action. Per the proposed RICT program, the RICT could be calculated to be any length of time between 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 30 days. Section 6.3.2 of the Vogtle FSAR states that ECCS components are designed such that a minimum of three accumulators, one charging pump, one safety injection pump, one residual heat removal pump, one Residual Heat Removal (RHR) heat exchanger, together with their associated valves and piping will ensure adequate core cooling in the event of a design basis accident.

The Vogtle TS Bases states that the need to ensure that three accumulators are adequate for this function is consistent with the loss-of-coolant accident (LOCA) assumption that the entire contents of one accumulator will be lost via the reactor coolant system (RCS) pipe break during the blowdown phase of the LOCA. It is not clear to the staff how the assumptions in the accident analysis would be satisfied for a LOCA in which the contents of one accumulator is lost through the break, and a second accumulator is inoperable at the time of the event. Please provide an explanation of how PRA functionality would be applied in this condition, why this condition would not be considered a TS loss of function, and how it would be assured that design basis success criteria would be satisfied.

RAI 10-5: The proposed changes to the TS include Condition 3.6.3.B, Containment Penetrations with more than one inoperable containment isolation valve, and Condition 3.6.3.C, Containment Penetrations with Purge Valves Leakage outside limits. The Required Action for Condition B is to isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. The current Completion Time to isolate the penetration flow path is one hour, which is consistent with the time specified to restore containment leakage to within its limits in TS LCO 3.6.1. Additionally, there is a requirement to verify the affected penetration flow path is isolated once per 31 days for devices outside containment.

Condition C applies when one or more penetration flow paths have one or more containment purge valves not within purge valve leakage limits. The required action is to isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. The proposed change is to permit the option of calculating a RICT for these Required Actions. Per the proposed RICT program, the RICT could be calculated to be any length of time between 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, for Condition B, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Condition C and 30 days. During this period, no actions would be required to isolate the affected penetration pathway(s);

and automatic actions to isolate the pathway may not be assured. The containment isolation valves form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on a containment isolation signal. The containment penetrations covered under conditions 3.6.3.B and C include those penetrations that are connected directly to the RCS or to the containment atmosphere, and are typically isolated using two isolation devices in series. If both of the isolation devices are inoperable in the open position, the safety function of minimizing the loss of reactor coolant inventory and maintaining the containment pressure boundary would not be assured. Please provide justification to support extension of the Completion Time up to a maximum of 30 days or remove these conditions from the scope of the RICT program. Please include an explanation of how PAA functionality would be applied in this Condition, why this condition would not be considered a TS loss of function, and how it would be assured that design basis success criteria would be satisfied.

RAI 11: In section 4.0 "Limitations and Conditions" of the NRC Staff safety evaluation (SE) to NEI 06-09, the staff stated: As part of its review and approval of a licensee's application requesting to implement the RMTS, the NRG staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods approved by the NRG staff for use in the plant-specific RMTS program. If a licensee wishes to change its methods, and the change is outside the bounds of the license condition, the licensee will need NRG approval, via a license amendment, of the implementation of the new method in its RMTS program. Please propose a license condition limiting the scope of the PAA and non-PAA methods to what is approved by the NRC staff for use in the plant-specific RMTS program. An example is provided below. The risk assessment approach, methods, and data shall be acceptable to the NRG, be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk from extending the completion times must be PRA methods accepted as part of this license amendment, or other methods currently approved by the NRG for generic use. If a licensee wishes to change its methods and the change is outside the bounds of this license condition, the licensee will need prior NRG approval, via a license amendment.

C. Pierce

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 -REQUEST FOR ADDITIONAL INFORMATION (CAC NOS. ME9555 AND ME9556) DATED MARCH 7, 2017 DISTRIBUTION:

PUBLIC LPL2-1 R/F RidsNrrDorlDpr Resource RidsACRS_MailCTR Resource RidsNrrLAKGoldstein Resource RidsNrrDorlLPL2-1 Resource RidsNrrPMVogtle Resource RidsRgn2MailCenter Resource ADAMS Accession No.: ML17034A084

• via email OFFICE DORL/LSPB/PM DORL/LPL2-1/LA DRA/APLA/BC DSS/STSB/BC NAME GEMiller KGoldstein SRosenberq*

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